Product consistency test of fully radioactive high-sodium content borosilicate glass K-26
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Product consistency test of fully radioactive high-sodium content borosilicate glass K-26 N.V. Ojovan1, I.V. Startceva1, A.S. Barinov1, M.I. Ojovan2, D.H. Bacon3, B.P. McGrail3, J.D. Vienna3 1
Scientific and Industrial Association “Radon”, Moscow, Russia.
2
Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, UK. 3
Pacific Northwest National Laboratory, Richland, Washington, US.
ABSTRACT Chemical durability of fully radioactive, high-sodium borosilicate glass K-26 was evaluated using the product consistency test PCT-A. Examination revealed normalised leaching rates as high as 5.93•10-2, 4.05•10-2 and 2.93•10-2 g/m2•day for sodium, boron and silicon respectively. Data on chemical durability of glass K-26 are consistent with similar composition glasses. These are of particular interest for performance assessment models. 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 durability and gave the basic possibility of maximal simplification of engineered barrier systems. The most simple disposal option is to emplace the waste form packages directly into earthen trenches provided the host rock has the necessary sorption and confinement properties. Such an approach has been implemented on an experimental scale as part of the research program of SIA ‘Radon’ on the immobilisation 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 and minor producers. Among the glasses used as waste forms for radionuclide immobilization, borosilicate glass is the most frequently used. The high sodium content borosilicate glass K-26 has been designed to immobilise intermediate level operational waste from nuclear power plants. Some tons of this glass were produced in the 1980s during a pilot vitrification campaign of radioactive waste from the Kursk nuclear power plant, which uses uranium-graphite channel type reactors. The glass K-26 had a density 2.46 g/cm3 with target composition on the oxide basis (wt.%): 48.2SiO2-7.5B2O3 - 2.5Al2O3 - 15.5CaO - 16.1Na2O 1.7Fe2O3 - 1.2NaCl - 1.1Na2SO4 - 6.2Misc. The main radioactive contaminant of waste was 137 Cs with the resulting concentration in the vitrified product as high as C0 =3.7•106 Bq/kg. The content of Na2O in the K-26 glass makes it relevant to the glass formulations being considered at Hanford site in the USA and hence data on
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