ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
Explore the many uses for nuclear science and its impact on energy, the environment, healthcare, food, and more.
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
Utility Working Conference and Vendor Technology Expo (UWC 2022)
August 7–10, 2022
Marco Island, FL|JW Marriott Marco Island
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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Nuclear Science and Engineering
September 2022
Nuclear Technology
Fusion Science and Technology
August 2022
Latest News
American Nuclear Society supports IAEA’s condemnation of attack on Ukraine nuclear power plant
Statement from American Nuclear Society President Steven Arndt and Executive Director and CEO Craig Piercy:
Kwang Soon Ha, Fan-Bill Cheung, Jinho Song, Rae Joon Park, Sang Baik Kim
Nuclear Technology | Volume 181 | Number 1 | January 2013 | Pages 196-207
Technical Paper | Special Issue on the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) / Thermal Hydraulics | dx.doi.org/10.13182/NT13-A15767
Articles are hosted by Taylor and Francis Online.
Boiling-induced natural-circulation flow in various engineered cooling channels is modeled and solved by considering the conservation of mass, momentum, and energy in the two-phase mixture, along with the two-phase friction drop and void fraction. The model is applied to estimate the induced mass flow rates through a uniform annular gap and a nonuniform annular gap between the reactor vessel and insulation under the in-vessel corium retention-external reactor vessel cooling conditions, and in the engineered corium cooling system of an ex-vessel core catcher during a severe accident. Dependence of the induced flow rate on various system parameters including the channel gap size, inlet diameter, inlet subcooling, and wall heat flux has been identified numerically. Results of the present study provide useful information for enhancing the design of engineered cooling channels to assure long-term cooling and retention of corium under severe accident conditions.