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Division Spotlight
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.
Meeting Spotlight
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver 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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Latest News
TerraPower begins U.K. regulatory approval process
Seattle-based TerraPower signaled its interest this week in building its Natrium small modular reactor in the United Kingdom, the company announced.
TerraPower sent a letter to the U.K.’s Department for Energy Security and Net Zero, formally establishing its intention to enter the U.K. generic design assessment (GDA) process. This is TerraPower’s first step in deployment of its Natrium technology—a 345-MW sodium fast reactor coupled with a molten salt energy storage unit—on the international stage.
Mario Dalle Donne, Ulrich Fischer, Marko Küchle
Nuclear Technology | Volume 71 | Number 1 | October 1985 | Pages 15-28
Technical Paper | Fission Reactor | doi.org/10.13182/NT85-A33707
Articles are hosted by Taylor and Francis Online.
A conceptual design of a helium-cooled blanket that satisfies the boundary conditions specified by the Next European Torus team is presented. The first wall is made of austenitic stainless steel with poloidally running helium cooling tubes and a 10-mm-thick steel erosion layer. The breeding material is lithium orthosilicate (Li4SiO4) with 60% 6Li enrichment and it is contained in the form of 2-mm pebbles in a bed together with 2-mm beryllium particles. Zirconium hydride is used in the back of the blanket to improve the tritium breeding. The main helium cooling system and the helium purge system for the tritium extraction are both at 80 bar, but they are completely separate for tritium-control reasons. An oxidizing atmosphere in the helium purge system ensures that the tritium losses from the plant are <10 Ci/day. The tritium inventory in the blanket is mainly due to tritium adsorption on the surface of the ceramic material. It is <1000 g, provided that the specific surface of the ceramic material is <0.25 m2/g. The rather leaky structure provided by the poloidally running breeder tubes is the main reason for the rather modest tritium breeding ratio. Improvement of the breeding ratio could be obtained by using a high melting point multiplier (beryllium or Zr5Pb3) in the first-wall region. This would also have the advantage of increasing the inlet helium temperature in the blanket region. The helium temperature resulting from the present design would allow a plant efficiency comparable to that of a pressurized water reactor. A higher plant efficiency would require the use of a more advanced structural material than austenitic stainless steel.
