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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
Meeting Spotlight
2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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Latest News
Glass strategy: Hanford’s enhanced waste glass program
The mission of the Department of Energy’s Office of River Protection (ORP) is to complete the safe cleanup of waste resulting from decades of nuclear weapons development. One of the most technologically challenging responsibilities is the safe disposition of approximately 56 million gallons of radioactive waste historically stored in 177 tanks at the Hanford Site in Washington state.
ORP has a clear incentive to reduce the overall mission duration and cost. One pathway is to develop and deploy innovative technical solutions that can advance baseline flow sheets toward higher efficiency operations while reducing identified risks without compromising safety. Vitrification is the baseline process that will convert both high-level and low-level radioactive waste at Hanford into a stable glass waste form for long-term storage and disposal.
Although vitrification is a mature technology, there are key areas where technology can further reduce operational risks, advance baseline processes to maximize waste throughput, and provide the underpinning to enhance operational flexibility; all steps in reducing mission duration and cost.
Jong-Dae Hong, Euijung Kim, Yong-Sik Yang, Dong-Hak Kook
Nuclear Technology | Volume 203 | Number 3 | September 2018 | Pages 282-292
Technical Paper | doi.org/10.1080/00295450.2018.1448203
Articles are hosted by Taylor and Francis Online.
One of the limiting mechanisms of pressurized water reactor spent fuel cladding is creep owing to high temperature and rod internal pressure. Based on extensive studies, many countries have tentatively concluded that creep rupture is hard to occur under dry storage conditions and cannot severely degrade the integrity of the cladding if it meets the 400°C limitation owing to a self-limiting property. However, the changes in mechanical properties after creep deformation are not well understood due to the limited amount of relevant tests and analyses. In this regard, mechanical property degradation of unirradiated Zircaloy-4 cladding by creep deformation was investigated using a ring compression test and microscopic observation. In addition, the implication regarding spent fuel cladding integrity based on the test results is described.