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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.
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2024 ANS Winter Conference and Expo
November 17–21, 2024
Orlando, FL|Renaissance Orlando at SeaWorld
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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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New laws offer nuclear industry incentives for existing power plant uprates
This year, the U.S. nuclear industry received a much-needed economic boost that could help preserve operating nuclear power plants and incentivize upgrades that extend their lifespan and power output.
Signed into law in 2022, the Inflation Reduction Act offers production tax credits (PTCs) for existing nuclear power plants and either PTCs or investment tax credits (ITCs) for new carbon-free generation. These credits could make power uprates—increasing the maximum power level at which a commercial plant may operate—a much more appealing option for utilities.
K. Yamaguchi, H. Nakamura, K. Haga
Nuclear Science and Engineering | Volume 88 | Number 3 | November 1984 | Pages 464-474
Technical Paper | doi.org/10.13182/NSE84-A18599
Articles are hosted by Taylor and Francis Online.
The effect of a local cooling disturbance caused by an edge-type blockage in a liquid-metal fast breeder reactor (LMFBR) fuel subassembly was investigated with a series of out-of-pile local blockage experiments with water and sodium. The heat exchange layer model first developed for central-type blockage cases applied well to the present edge-type cases. An empirical formula was developed for estimating maximum temperatures in various subassemblies, and the conclusion was reached that a middle size edge-type blockage could lead to sodium boiling. The critical heat flux data of Power Reactor and Nuclear Fuel Development Corporation and Kernforschungszentrum Karlsruhe were correlated with the boiling inception heat flux for various core flow velocities. A linear relation was found between them, suggesting a possible interpretation of the coolability limit within the framework of nonboiling conditions. The theoretical (hypothetical) excess temperature in the absence of boiling, ΔTms (= Tmax − Tsat), seemed to cross a critical value at the instance of permanent dryout. Based on the constant critical ΔTms assumption and the formula for Tmax, an assessment was made of the thermohydraulic consequences for the different blockage size situations of a typical LMFBR.