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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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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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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General Kenneth Nichols and the Manhattan Project
Nichols
The Oak Ridger has published the latest in a series of articles about General Kenneth D. Nichols, the Manhattan Project, and the 1954 Atomic Energy Act. The series has been produced by Nichols’ grandniece Barbara Rogers Scollin and Oak Ridge (Tenn.) city historian David Ray Smith. Gen. Nichols (1907–2000) was the district engineer for the Manhattan Engineer District during the Manhattan Project.
As Smith and Scollin explain, Nichols “had supervision of the research and development connected with, and the design, construction, and operation of, all plants required to produce plutonium-239 and uranium-235, including the construction of the towns of Oak Ridge, Tennessee, and Richland, Washington. The responsibility of his position was massive as he oversaw a workforce of both military and civilian personnel of approximately 125,000; his Oak Ridge office became the center of the wartime atomic energy’s activities.”
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 | 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.