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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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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Latest News
Commercial nuclear innovation "new space" age
In early 2006, a start-up company launched a small rocket from a tiny island in the Pacific. It exploded, showering the island with debris. A year later, a second launch attempt sent a rocket to space but failed to make orbit, burning up in the atmosphere. Another year brought a third attempt—and a third failure. The following month, in September 2008, the company used the last of its funds to launch a fourth rocket. It reached orbit, making history as the first privately funded liquid-fueled rocket to do so.
Tadas Kaliatka, Eugenijus Uspuras, Algirdas Kaliatka
Fusion Science and Technology | Volume 72 | Number 2 | August 2017 | Pages 176-187
Technical Note | doi.org/10.1080/15361055.2017.1320496
Articles are hosted by Taylor and Francis Online.
An event of water coolant ingress into the vacuum vessel (VV) is one of the most important events leading to severe consequences in nuclear fusion reactors. The ingress of coolant to the VV could appear due to coolant pipe rupture of in-vessel components. Any damage of in-vessel components could lead to water ingress and may lead to pressure increase and possible damage of the VV. Therefore, it is important to understand thermohydraulic processes in the VV during the ingress of coolant event (ICE) to prevent overpressurization of the VV. This technical note updates the developed Wendelstein 7-X (W7-X) model in accordance with the experience gained from the modeling of ICE experiments. Calculation results using the updated model are compared with the results obtained using an older model and the results of other researchers. The calculation results of the updated W7-X model show a much smaller pressure increase rate in the VV compared to the old model. In order to find the maximal area of partial break, which increases pressure in the VV but does not reach burst disk activation pressure (no steam release from the VV to the environment), the best-estimate approach is provided. The results of the analysis reveal that partial break using the updated W7-X model could be much bigger than what was considered before.