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Division Spotlight
Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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.
W. G. Schuetzenduebel
Nuclear Technology | Volume 28 | Number 3 | March 1976 | Pages 315-327
Technical Paper | Reactor | doi.org/10.13182/NT76-A31514
Articles are hosted by Taylor and Francis Online.
Advances in steam generator design have been made in recent years. The demands of gas-cooled nuclear power plants mean high-temperature operating conditions and space limitations. The feasibility of the high-temperature gascooled reactor (HTGR) concept and the 235U-Th233U fuel cycle was demonstrated by 6 yr of operation of the 40-MW(e) Peach Bottom prototype HTGR power plant. Two steam generators located outside the pressure vessel were used to exchange the heat from the primary coolant (helium) to the secondary coolant (water). A prestressed concrete reactor vessel (PCRV) was used in the design of the 330-MW(e) Fort St. Vrain power demonstration plant. Use of the PCRV made the integration of all the nuclear steam supply system components practical. The primary coolant inventory was reduced and external piping and steam generator pressure shells were eliminated. A once-through-type steam generator system was selected. Materials selected for use in the pressure parts exceeded American Society of Mechanical Engineers Code requirements. The next step in the development of HTGR technology is the large commercial HTGR plant, which has once-through-type steam generators with a nominal capacity of 500 MW(th). Materials used in the main steam section range from 2¼ Cr—1 Mo to Ni-Fe-Cr (Alloy 800). High carbon levels were used to increase the creep strength of the materials. Gas cooling for fast breeder reactors is being studied by designing a 300-MW(e) demonstration plant. The steam generators are similar to the design of the Fort St. Vrain and large commercial plants. Tubes made of Alloy 800 are used.
