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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
2023 ANS Annual Meeting
June 11–14, 2023
Indianapolis, IN|Marriott Indianapolis 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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Nuclear Science and Engineering
June 2023
Nuclear Technology
July 2023
Fusion Science and Technology
Latest News
Destruction of Ukrainian dam threatens Zaporizhzhia
A Soviet-era dam downstream from the Zaporizhzhia nuclear power plant in southeastern Ukraine collapsed last evening, causing the water level of the Kakhovka Reservoir north of the dam to drop and raising new concerns over the already jeopardized safety of the Russian-occupied nuclear facility, Europe’s largest. The reservoir supplies water for, among other things, Zaporizhzhia’s cooling systems.
Technical Session
Friday, April 9, 2021|11:00AM–12:30PM EDT
Session Chair:
Yuqiao Joy Fan (NCSU)
Alternate Chair:
Ishita Trivedi
Session Organizer:
Edward Chen (NCSU)
Track Organizer:
Session Producers:
Halil Ibrahim San (NCSU)
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Pressure Distribution in Free and Impinging Jets for High-Energy Piping Systems
Zhengting Quan (Purdue University), Ran Kong (Purdue University), Seungjin Kim (Purdue University), Mamoru Ishii (Purdue University)
Summary
Thermal Decomposition of Carbonates as a Passive Cooling Mechanism for Molten Salt Reactors
Christopher T. Forsyth (University of California Berkeley)
Improvements in Bypass flow Analysis Using CORONA Code
Dylan D. Ohrt (University of Texas Permian Basin)
Design of a Sodium Heat Pipe Experimental Setup for the Special Purpose Nuclear Reactor
Pei-Hsun Huang (University of Michigan), Taehwan Ahn (University of Michigan), Annalisa Manera (University of Michigan), Victor Petrov (University of Michigan)
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