Abstract
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: 93mNb and the long-lived 94Nb, 93Zr radionuclides. Radionuclides of 60Co, 137Cs, 90Sr, 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 60Co and 137Cs 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.
Highlights
The decommissioning of uranium-graphite reactors (UGR) presents a wide range of issues that demand the solutions both in the preparation phase and directly during the process of decommissioning
The technological channels (TC) pipes of zirconium alloy were used in the final stages of operation of some production uranium-graphite reactors (PUGR); the pipes were installed in certain zones of the graphite stack to maintain its integrity
Conclusions results of the accomplished studies allowed to certify the following: 1) The dynamics of leaching and the absolute leach rates values of 93mNb, 94Nb, 93Zr, 60Co, 137Cs, 90Sr, and actinide radionuclides identified in the samples of the irradiated technological channels of the E110 zirconium alloy witnessed the high corrosion resistance of the irradiated structural elements and parts of the zirconium alloy
Summary
Series: Materials Science and Engineering 135 (2016) 012020 doi:10.1088/1757-899X/135/1/012020. S G Kotlyarevskiy, A O Pavliuk, E V Zakharova and A G Volkova Pilot & Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors, Seversk, Tomsk region, Russia 2 Tomsk polytechnic university, Tomsk, Russia 3 Frumkin IPCE RAS, Moscow, Russia
Sign-in/Register to view highlights & summary Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callMore From: IOP Conference Series: Materials Science and Engineering
Disclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
