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Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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Fusion Science and Technology
Latest News
A more open future for nuclear research
A growing number of institutional, national, and funder mandates are requiring researchers to make their published work immediately publicly accessible, through either open repositories or open access (OA) publications. In addition, both private and public funders are developing policies, such as those from the Office of Science and Technology Policy and the European Commission, that ask researchers to make publicly available at the time of publication as much of their underlying data and other materials as possible. These, combined with movement in the scientific community toward embracing open science principles (seen, for example, in the dramatic rise of preprint servers like arXiv), demonstrate a need for a different kind of publishing outlet.
E. A. Mogahed
Fusion Science and Technology | Volume 44 | Number 1 | July 2003 | Pages 69-73
Technical Paper | Fusion Energy - MFE Chamber Technology | doi.org/10.13182/FST03-A312
Articles are hosted by Taylor and Francis Online.
A helical coolant channel scheme is proposed for the APEX solid wall blanket module. The self-coolant breeder in this system is FLIBE (LiF)2(BeF2). The structural material is the nanocomposited alloy 12YWT. The neutron multiplier used in the current design is either stationary or slow moving liquid lead. The purpose of this study is to design a blanket that can handle a high wall loading (5 MW/m2). In the mean time the design provides means to attain the maximum possible blanket outlet temperature and meet all engineering limits on temperature of structural material and liquids. An important issue for such a design is to optimize the system for minimum pressure loss. For advanced ferritic steel (12YWT) an upper temperature limit of 800°C is expected, and a limit of 700°C at the steel/FLIBE interface is recommended.The blanket module is composed of two main continuous routes. The first route is three helical rectangular channels side-by-side that surround a central box. The helical channels are fed from the bottom and exit at the top to feed the central channels in the central box. The coolant helical channels have a cross sectional area with a length of about 10 cm and a width that changes according to the position around the central box. For instance: the width of the coolant channels facing the plasma is the narrowest while it is the widest in the back (farthest from the plasma).In this design the coolant runs around the central box for only 5 turns to cover the total height of the first wall (6.8 m). The design is optimized with the FW channel width as a parameter with the heat transfer requirements at the first wall as the constraints.
