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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
2024 ANS Winter Conference and Expo
November 17–21, 2024
Orlando, FL|Renaissance Orlando at SeaWorld
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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December 2024
Nuclear Technology
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November 2024
Latest News
Diablo Canyon completes dry storage campaign, seeks ISFSI license renewal
Holtec International announced that it has completed the campaign to transfer Diablo Canyon’s spent nuclear to dry storage ahead of its planned schedule, paving the way for the continued operation of the central California nuclear power plant.
Alice Ying, Haibo Liu, Mohamed Abdou
Fusion Science and Technology | Volume 64 | Number 2 | August 2013 | Pages 303-308
Divertor and High-Heat-Flux Components | Proceedings of the Twentieth Topical Meeting on the Technology of Fusion Energy (TOFE-2012) (Part 1), Nashville, Tennessee, August 27-31, 2012 | doi.org/10.13182/FST64-303
Articles are hosted by Taylor and Francis Online.
Available data and mathematical formulations concerning tritium transport in the FW/Divertor with tungsten and beryllium as plasma facing materials were implemented in the commercial code COMSOL Multiphysics. The goal is to develop a CAD-based multiphysics modeling capability so that FW/Divertor temperature and geometric features can be readily taken into consideration while tritium permeation to the primary coolant in a prototypical PFC can be more realistically addressed. This development began with the simulation of ion implantation experiments, validated against existing laboratory experimental results. Analysis shows that with ITER FW where Be is used as the plasma facing material, the low operating temperature, erosion, and the dwell time greatly hinder tritium bulk diffusion, permeation, and inventory accumulation. However, under DEMO high-temperature operating conditions, tritium can quickly diffuse through tungsten to structural material and reach a steady state inventory after a relatively short time. Additionally, its permeation to the coolant can be reduced when the Soret effect is considered. The findings and challenges of developing a 3-D predictive capability for tritium transport in a FW/Divertor PFC are discussed.