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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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Powering the future: How the DOE is fueling nuclear fuel cycle research and development
As global interest in nuclear energy surges, the United States must remain at the forefront of research and development to ensure national energy security, advance nuclear technologies, and promote international cooperation on safety and nonproliferation. A crucial step in achieving this is analyzing how funding and resources are allocated to better understand how to direct future research and development. The Department of Energy has spearheaded this effort by funding hundreds of research projects across the country through the Nuclear Energy University Program (NEUP). This initiative has empowered dozens of universities to collaborate toward a nuclear-friendly future.
Shangzhen Xie, Jiyun Zhao (City Univ of Hong Kong)
Proceedings | 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) | Charlotte, NC, April 8-11, 2018 | Pages 172-181
External reactor vessel cooling (ERVC) is proved as a necessary system in in-vessel retention management in the nuclear power plant to transfer the heat from failed core to outside vessel, in the aim of preserving intact vessel and avoiding severe accidents. To provide advanced safety guarantee for the next generation of nuclear power plant, a greater designed safety margin should be considered and proposed, such as increasing the tolerance of high heat flux by using advanced materials of the vessel, insulated structures between reactor core and the vessel, and superior coolant in ERVC system. As long as the heat flux of the reactor vessel wall emerged from melt-core does not go beyond the maximum limitation?Critical Heat Flux (CHF), the decay heat can be dissipated timely and thus the reactor can be cooled down without releasing radiation products. In this case, increasing critical heat flux by various approaches is deemed essential also attract intensive studies in nuclear systems. In fact, the research of the enhancement of critical heat flux has a long history, with extensive experiments and simulations devoted in the last several decades to seeking for methods to expand thermal margin and it continues to be a promising topic in heat transfer research fields. In this paper, we present a comprehensive overview of CHF enhancement experiments, focusing on four broad categories of approaches. The first approach considered is amelioration of fluid properties by adding nanoparticles into the base fluid, by which both flow boiling and pool boiling achieve significant improvements in CHF. The second prevailing method recently is surface modifications by various advanced techniques. Third, we review the effect of various modified channel structures on the boiling process. Finally, some creative and notable hybrid techniques are presented. Based on this review of the state-of-the-art in CHF enhancement, future research directions are also proposed.