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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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Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
J. E. Chafey, D. I. Roberts
Nuclear Technology | Volume 55 | Number 1 | October 1981 | Pages 37-49
Technical Paper | Materials Performance in Nuclear Steam Generator / Material | doi.org/10.13182/NT81-A32830
Articles are hosted by Taylor and Francis Online.
High temperature gas-cooled reactor (HTGR) systems feature a graphite-moderated, uranium-thorium, all-ceramic core and utilize high pressure helium as the primary coolant. The steam generators in these systems are exposed to gas-side temperatures approaching 760°C (1400°F) and produce superheated steam at 538°C (1000°F) and 16.5 MPa (2400 psi). Thus, the design and development of steam generators for these systems require consideration of time-dependent materials behavior, corrosion, fretting, wear, and other related phenomena of concern in all steam generators. The prototype Peach Bottom Unit No. 1 40-MW (electric) HTGR was operated by the Philadelphia Electric Company for a total of 1349 equivalent full power days during a 7-yr period. Upon planned decommissioning of that plant, the forced-recirculation U-tube steam generators and other components were subjected to destructive properties tests and metallurgical examinations. These tests and examinations showed the steam generators to be in very satisfactory condition. The 330-MW(electric) Fort St. Vrain HTGR, owned and operated by Public Service Company of Colorado, and now in the final stages of startup, has achieved 70% power and generated more than 1.5 × 106 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, and their design and development required considering a number of new materials factors including creep fatigue. Also, because of the once-through design, water chemistry control needed special consideration. Current designs of larger HTGRs also feature steam generators of helical tube once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of very large tubesheet forgings, consideration of the gaseous corrosion effects of the primary coolant, and other related factors
