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Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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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
Hsoung-Wei Chou, Pin-Chiun Huang, Yuh-Ming Ferng
Nuclear Technology | Volume 207 | Number 5 | May 2021 | Pages 735-749
Technical Paper | doi.org/10.1080/00295450.2020.1796091
Articles are hosted by Taylor and Francis Online.
The structural integrity of a reactor pressure vessel (RPV) is a crucial issue for an operating nuclear power plant, especially in the beltline region, which suffers the highest neutron irradiation. Owing to its capability of considering parameters based on statistical distributions and provision of objective risk-informed results, the probabilistic fracture mechanics (PFM) method is widely used in evaluating the structural integrity of RPVs. However, the flaw characteristics used for PFM analysis are mainly derived from particular vessel inspection information such as from the Pressure Vessel Research User Facility and Shoreham vessels, which may not be able to truly represent the vessel-specific condition of an analyzed RPV. In this work, the Bayesian inference, which combines prior flaw data with new inspection results as well as uncertainties, is used to develop posterior vessel-specific flaw distributions. Then, the updated flaw model is used for PFM analysis to investigate the effects on the fracture probability assessment of RPVs subjected to pressurized thermal shocks (PTSs). Considering the updated flaws based on the inspection data, the PFM analysis result could be more realistic to predict the fracture risks of RPVs during operation.
