Corrosion mechanisms of low level vitrified radioactive waste in a loamy soil
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Corrosion mechanisms of low level vitrified radioactive waste in a loamy soil M.I. Ojovan1, W.E. Lee1, A.S. Barinov2, N.V. Ojovan2, I.V. Startceva2, D.H. Bacon3, B.P. McGrail3, J.D. Vienna3 1
Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, UK; 2 Scientific and Industrial Association “Radon”, Moscow, Russia; 3 Pacific Northwest National Laboratory, Richland, Washington, US. ABSTRACT Field experiments have run for over 14 years to evaluate the behaviour of the same high-sodium content radioactive waste borosilicate glass buried in a loamy soil (glass K-26) and in an open testing area (glass Bs-10). Processing of field data for glass Bs-10 tested in an open area has resulted in a dissolution rate r = 0.42 µm/y and caesium diffusion coefficient D ≈ 1.8 10-20 m2/s at testing temperatures up to 19 oC. Both ion-exchange and hydrolysis control the corrosion of this glass. Processing of field data for K-26 glass revealed an insignificant role of glass dissolution. The caesium diffusion coefficient was estimated as D ≈ (3.4-5.1) 10-21 m2/s. Due to the relatively low storage temperatures (4.5 oC) used the leaching behaviour of glass K-26 is believed to be controlled by ion exchange processes. This mechanism is likely to remain dominant until the decay of 137Cs in the glass is below exemption levels. INTRODUCTION Vitrification of low and intermediate level radioactive waste (LILW) is attracting great interest and large programmes are currently underway or planned in a number of countries including Russia and the USA. Utilisation of glass as a waste form for LILW requires appropriate performance assessment support for waste disposal facilities, which presumes an understanding of the main glass corrosion mechanisms. The corrosion of LILW glass has not been as intensively studied as that of high-level vitrified waste. Studies of actual vitrified radioactive waste in conditions similar to those expected in the disposal environment give realistic data diminishing many of the uncertainties, and enabling validation of computer code simulation to improve the confidence of mathematical models [1]. This paper examines the corrosion behaviour of identical radioactive high-sodium content borosilicate glass in a loamy soil (glass K-26) and in an open testing area (glass Bs-10). The borosilicate glass K-26 (Bs-10) has been designed to immobilise intermediate level operational nuclear power plant (NPP) radioactive waste. Some tonnes of this glass were produced in the 1980-s using radioactive waste from the Kursk NPP in Russia (with a channel type reactor RBMK) and a number of glass blocks have been disposed of in an experimental shallow land facility for long-term corrosion tests. K-26 waste glass has a density 2.46 g/cm3 and theoretical composition on an oxide basis of (wt.%): 48.2 SiO2-7.5 B2O3 2.5 Al2O3 15.5 CaO 16.1 Na2O 1.7 Fe2O3 1.2 NaCl 1.1 Na2SO4 6.2Misc [2]. The Na2O content in K26 glass (16.1) makes it relevant to the glass formulations being considered at the Hanford site i
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