On the evaluation of absorbed dose in vitrified high level radioactive waste with the account of real can geometry

Authors:   Aloy A. S., Kovalev N. V., Prokoshin A. M., Blokhin A. I., Blokhin P. A., Kuryndin A. V., Ponizov A. V., Makovsky S. V. Issue:   4 (98) – 2020.

A large amount of radioactive waste is produced during reprocessing of spent nuclear fuel. Most of this radioactive waste is subject to vitrification in glass-like compound and to subsequent deep geological disposal. One of the main requirements to radioactive waste compound is radiological stability under radiation of nuclides in waste for a long period of time. An amount of absorbed dose in compound volume shall be determined in order to justify radiological stability of radioactive waste. This article presents the results of absorbed dose calculation in radioactive waste in borsilicate glass matrix during long-term storage and subsequent disposal.

Keywords: radioactive waste, calculation modeling, absorbed dose.

Article language: Russian. Pp. 61–72. DOI: 10.26277/SECNRS.2021.98.4.006.


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