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Nuclear Installations Safety
Devoted specifically to the safety of nuclear installations and the health and safety of the public, this division seeks a better understanding of the role of safety in the design, construction and operation of nuclear installation facilities. The division also promotes engineering and scientific technology advancement associated with the safety of such facilities.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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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Nuclear News 40 Under 40 discuss the future of nuclear
Seven members of the inaugural Nuclear News 40 Under 40 came together on March 4 to discuss the current state of nuclear energy and what the future might hold for science, industry, and the public in terms of nuclear development.
To hear more insights from this talented group of young professionals, watch the “40 Under 40 Roundtable: Perspectives from Nuclear’s Rising Stars” on the ANS website.
Thiago D. Roberto, Celso M. F. Lapa, Antonio C. M. Alvim
Nuclear Technology | Volume 206 | Number 4 | April 2020 | Pages 527-543
Technical Paper | doi.org/10.1080/00295450.2019.1666603
Articles are hosted by Taylor and Francis Online.
Reactor cavity cooling systems (RCCSs) ensure the physical integrity of the containment structures in a high-temperature gas-cooled test reactor (HTR-10) and a high-temperature gas-cooled pebble-bed module reactor (HTR-PM). HTR-10 is a graphite-moderated and helium-cooled pebble-bed reactor prototype designed to demonstrate the technical feasibility and safety of the pebble-bed reactor design concept under normal and accident conditions. This prototype served as a proof of concept for the HTR-PM that shares several design similarities with the HTR-10, including a reactor cavity that requires cooling owing to the high core outlet temperature. The RCCS conceived in the design of both the reactors increases the inherent safety of the system by dissipating heat through passive heat removal processes. This paper proposes an RCCS model for system-scale analysis. The conventional scale method is adopted to determine the conditions necessary for complete similarity between two RCCSs in the steady-state flow regime. In addition, a scaling evaluation between the RCCSs of both the HTR-10 (model) and HTR-PM (prototype) is performed using the proposed RCCS model based on data from two benchmark problems: pressurized and depressurized loss of forced cooling. This evaluation shows that the RCCSs of the HTR-10 (model) and HTR-PM (prototype) show similarity to a specific operational condition in each of the problems analyzed.