ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
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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
2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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
Smarter waste strategies: Helping deliver on the promise of advanced nuclear
At COP28, held in Dubai in 2023, a clear consensus emerged: Nuclear energy must be a cornerstone of the global clean energy transition. With electricity demand projected to soar as we decarbonize not just power but also industry, transport, and heat, the case for new nuclear is compelling. More than 20 countries committed to tripling global nuclear capacity by 2050. In the United States alone, the Department of Energy forecasts that the country’s current nuclear capacity could more than triple, adding 200 GW of new nuclear to the existing 95 GW by mid-century.
Yoshiyuki Inagaki, Kazuhiko Kunitomi, Yoshiaki Miyamoto, Ikuo Ioka, Kunihiko Suzuki
Nuclear Technology | Volume 99 | Number 1 | July 1992 | Pages 90-103
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT92-A34706
Articles are hosted by Taylor and Francis Online.
The high-temperature engineering test reactor (HTTR) is a 30-MW(thermal) helium gas-cooled reactor being constructed by the Japan Atomic Energy Research Establishment. A thermal mixing study of the coolant in the core bottom structure (CBS) of the HTTR is conducted to clarify the thermal-hydraulic characteristics of the coolant and estimate the influence of a hot streak on the intermediate heat exchanger (IHX) and a pressurized water cooler (PWC) downstream from the core. An experiment is carried out using an in-core structure test section (a full-scale simulation model of the CBS) of the helium engineering demonstration loop (HENDEL), and a numerical analysis is made using a three-dimensional time-dependent flow and heat transfer code including a k-ε model of turbulence. It is confirmed that the coolant is mixed sufficiently in the CBS and the outlet gas duct of the HTTR, and the hot streak had little effect on the IHX and the PWC.
