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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.
Kazuhiro Kobayashi, Yuji Torikai, Makiko Saito, Vladimir Alimov, Naoyuki Miya, Yoshitaka Ikeda
Fusion Science and Technology | Volume 67 | Number 2 | March 2015 | Pages 428-431
Proceedings of TRITIUM 2013 | doi.org/10.13182/FST14-T46
Articles are hosted by Taylor and Francis Online.
Disassembly of the JT-60U torus was started in 2010 after 18 years D2 operations. In future the vacuum vessel will be treated as non-radioactive ones after the clearance procedure under the Japanese regulation depending on the tritium (T) contamination level. Note that the vessel was manufactured from Inconel 625 steel. Therefore, it was very important to study the hydrogen isotope behavior in Inconel 625 from viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D2 (92.8 %) – T2 (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily. From these results, the behavior of tritium in the vacuum vessel of the JT-60U torus will be evaluated and discussed