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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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Latest News
First astatine-labeled compound shipped in the U.S.
The Department of Energy’s National Isotope Development Center (NIDC) on March 31 announced the successful long-distance shipment in the United States of a biologically active compound labeled with the medical radioisotope astatine-211 (At-211). Because previous shipments have included only the “bare” isotope, the NIDC has described the development as “unleashing medical innovation.”
Bernard R. Bandini, Kostadin N. Ivanov, Anthony J. Baratta, Robert G. Steinke
Nuclear Technology | Volume 123 | Number 1 | July 1998 | Pages 1-20
Technical Paper | Reactor Safety | doi.org/10.13182/NT98-A2875
Articles are hosted by Taylor and Francis Online.
The verification of a three-dimensional nodal transient neutronics routine in the TRAC-PF1/MOD3 Version 1.0 thermal-hydraulic system analysis computer code is discussed. This neutronics algorithm is based on a fully implicit transient version of the well-known nodal expansion method. Results from running TRAC-PF1/MOD3 with this new neutronics routine were compared with the results of running two established neutronics/thermal-hydraulic space-time codes, HERMITE and ARROTTA. The transient chosen for this code verification was a rapid ejection of an off-center control rod in a Westinghouse pressurized water reactor, which is initially at hot standby. This severe prompt-critical transient provides a stringent test of TRAC-PF1/MOD3's new multidimensional neutronics routine and its coupling to the existing thermal-hydraulic solution methodology. Because of its speed, the transient tests only the fuel rod heat conduction coupling and not the coolant thermal-hydraulic coupling.Acceptable agreement was obtained among the results from TRAC-PF1/MOD3, HERMITE, and ARROTTA during all phases of this transient. Agreement was in the areas of time dependence of total-core and peak-assembly powers, as well as the time dependence of the core-average and peak-assembly fuel temperatures. In addition, comparison of several steady-state calculations that provide initial conditions for the transient analysis showed acceptable agreement in the calculated eigenvalues and normalized assembly-power distributions.