Пожалуйста, используйте этот идентификатор, чтобы цитировать или ссылаться на этот ресурс: http://earchive.tpu.ru/handle/11683/34809
Название: Capability assessment for application of clay mixture as barrier material for irradiated zirconium alloy structure elements long-term processing for storage during decommissioning of uranium-graphite nuclear reactors
Авторы: Kotlyarevskiy, S. G.
Pavliuk, A. O.
Zakharova, E. V.
Volkova, A. G.
Ключевые слова: глиняные массы; барьерные материалы; сплавы циркония; уран-графитовые ядерные реакторы; радионуклиды; глины; облученные сплавы
Дата публикации: 2016
Издатель: IOP Publishing
Библиографическое описание: Capability assessment for application of clay mixture as barrier material for irradiated zirconium alloy structure elements long-term processing for storage during decommissioning of uranium-graphite nuclear reactors / S. G. Kotlyarevskiy [et al.] // IOP Conference Series: Materials Science and Engineering. — 2016. — Vol. 135 : Issues of Physics and Technology in Science, Industry and Medicine : VIII International Scientific Conference, 1–3 June 2016, Tomsk, Russia : [proceedings]. — [012020, 7 p.].
Аннотация: The radionuclide composition and the activity level of the irradiated zirconium alloy E110, the radionuclide immobilization strength and the retention properties of the mixed clay barrier material with respect to the radionuclides identified in the alloy were investigated to perform the safety assessment of handling structural units of zirconium alloy used for the technological channels in uranium-graphite reactors. The irradiated zirconium alloy waste contained the following activation products:{93m}Nb and the long-lived {94}Nb, {93}Zr radionuclides. Radionuclides of {60}Co, {137}Cs, {90}Sr, and actinides were also present in the alloy. In the course of the runs no leaching of niobium and zirconium isotopes from the E110 alloy was detected. Leach rates were observed merely for {60}Co and {137}Cs present in the deposits formed on the internal surface of technological channels. The radionuclides present were effectively adsorbed by the barrier material. To ensure the localization of radionuclides in case of the radionuclide migration from the irradiated zirconium alloy into the barrier material, the sorption properties were determined of the barrier material used for creating the long-term storage point for the graphite stack from uranium-graphite reactors.
URI: http://earchive.tpu.ru/handle/11683/34809
Располагается в коллекциях:Материалы конференций

Файлы этого ресурса:
Файл РазмерФормат 
dx.doi.org-10.1088-1757-899X-135-1-012020.pdf1,12 MBAdobe PDFПросмотреть/Открыть


Все ресурсы в архиве электронных ресурсов защищены авторским правом, все права сохранены.