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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
ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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
Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
Technical Session|Thermal Hydraulics
Friday, April 5, 2024|3:15–3:45PM EDT|Reber Room 135
Session Chair:
David Reger
Alternate Chair:
John J. Acierno
Session Organizer:
Carolina da Silva Bourdot Dutra (Penn State)
Dr. Joshua Hansel (INL) will present on Sockeye, a heat pipe analysis application based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element framework, specifically designed for modeling heat pipes in heat-pipe-cooled nuclear microreactors. This talk will cover Sockeye's transient heat pipe modeling capabilities, validation using data from the Michigan Single Sodium Heat Pipe (MISOH1) facility, and the importance of higher fidelity heat pipe models and accurate knowledge of internal dimensions for producing realistic transient behavior. Dr. Hansel will emphasize Sockeye's role in advancing the understanding and design of heat-pipe-cooled nuclear microreactors and discuss future development plans. *This presentation will be 30 minutes instead of 20 minutes.*
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A Sockeye Model of a Test at the Michigan Single Sodium Heat Pipe Facility
3:15–3:45PM EDT
Joshua E. Hansel (INL), Carolina da Silva B. Dutra (Penn State), Pei-Hsun Huang (Univ. Michigan), Taehwan Ahn (Univ. Michigan), Victor Petrov (Univ. Michigan)
Paper
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