Simulation of the Waste Glass Behavior in a Loamy Soil of the Wet Repository Site
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Simulation of the Waste Glass Behavior in a Loamy Soil of the Wet Repository Site M.I. Ojovan, N.V. Ojovan, I.V. Startceva, G.N. Chuikova and A.S. Barinov Scientific and Industrial Association ‘Radon’, the 7-th Rostovsky Lane 2/14, Moscow, Russia Fax: (095) 248 1941, E-mail: [email protected] ABSTRACT A model developed for description of waste glass corrosion has been applied to assess the radionuclide release from real radioactive (intermediate level) vitrified material over extended storage periods. Field data generated during the long-term testing of the prototype waste glass packages were mathematically processed and the derived parameters used in model calculations. Regardless of the corrosive saturated conditions of the wet near-surface repository, the fairly high safety of trench disposal has been demonstrated for borosilicate glass containing real NPPoperational waste. INTRODUCTION Waste forms disposed of in a near-surface wet repository eventually come into contact with groundwater. Engineered structures used or designed to prevent or postpone such contact and the subsequent radionuclide release are complex and often too expensive. Development of vitrification technologies by the beginning of the 1970s provided waste forms with excellent resistance to corrosion and gave the basic possibility of maximal simplification of engineered barrier systems. The most simple disposal option is to locate the waste form packages directly into earthen trenches provided the host rock has the necessary sorbing and confinement properties. Such an approach has been implemented on an experimental scale as part of the research program of SIA ‘Radon’ on the encapsulation of radioactive waste in a glass matrix. The program was initiated in the mid-70s. Since then pilot and industrial vitrification plants have been constructed based on the use of a ceramic Joule-heated melter, plasma melter, and induction cold-crucible melting process. Vitrification techniques were applied to liquid and solid LILW received by the site from various sources including the Moscow wastewater purification plant, nuclear power plants (NPP) and minor producers. Among the glasses used as waste forms for radionuclide immobilization, borosilicate (BS) glass is the most frequently used. Packages of BS waste glass in carbon steel canisters containing intermediate level operational waste from NPP were manufactured and placed for testing into experimental near-surface repositories and in an open site. This paper describes the results of processing the data collected during the 12-yr monitoring of waters contacting the waste forms and aims to predict waste form leaching behavior over extended time periods.
EXPERIMENT Real medium level operational waste from the Kursk NPP was used. The waste is sludge with a salt content of about 340 g/l and NaNO3 as the main waste component (86 wt%). The main radionuclides are 137Cs (82%), 134Cs, and 90Sr. The molten borosilicate waste glass was produced in the pilot plant «EPOS» with a slurry-fed ceramic Joule-heated glass melter a
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