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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.
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2025 ANS Annual Conference
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Chicago, IL|Chicago Marriott Downtown
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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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High-temperature plumbing and advanced reactors
The use of nuclear fission power and its role in impacting climate change is hotly debated. Fission advocates argue that short-term solutions would involve the rapid deployment of Gen III+ nuclear reactors, like Vogtle-3 and -4, while long-term climate change impact would rely on the creation and implementation of Gen IV reactors, “inherently safe” reactors that use passive laws of physics and chemistry rather than active controls such as valves and pumps to operate safely. While Gen IV reactors vary in many ways, one thing unites nearly all of them: the use of exotic, high-temperature coolants. These fluids, like molten salts and liquid metals, can enable reactor engineers to design much safer nuclear reactors—ultimately because the boiling point of each fluid is extremely high. Fluids that remain liquid over large temperature ranges can provide good heat transfer through many demanding conditions, all with minimal pressurization. Although the most apparent use for these fluids is advanced fission power, they have the potential to be applied to other power generation sources such as fusion, thermal storage, solar, or high-temperature process heat.1–3
Belal Almomani, Ahmad Ababneh, Muhammad Zubair
Nuclear Technology | Volume 209 | Number 2 | February 2023 | Pages 214-227
Technical Paper | doi.org/10.1080/00295450.2022.2133507
Articles are hosted by Taylor and Francis Online.
After the Three Mile Island Unit 2 accident, regulatory bodies were concerned about the safety measures in design and operation corresponding to the operator’s decisions and procedures for handling such off-normal transients. Several recommendations were proposed to analyze transients and accidents, improve and revise emergency operating procedures (EOPs), and conduct functional training. In this work, procedural paths were systematically studied to identify the problems in the diagnosis associated with a pilot-operated relief valve (PORV)–break loss-of-coolant accident (LOCA) as well as to suggest new indications for improving the EOPs. Operational parameters during PORV-break LOCA and pipeline-break small-break LOCA were analyzed using a generic pressurized water reactor simulator to compare and justify the symptoms between these two events. It was found that suggesting further indications mainly in the reactor cooling system and containment symptoms may improve the diagnosis of a PORV-break LOCA from the pipeline-break small-break LOCA. This paper presents a practical approach to evaluating diagnostic procedures to better understand operator recovery actions corresponding to reactor system response in dealing with a PORV-break LOCA.
