Study of BiPbO 2 NO 3 for I-129 Fixation under Reducing Conditions
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Study of BiPbO2NO3 for I-129 Fixation under Reducing Conditions Takayuki Amaya1, Atsushi Mukunoki1, Mamoru Shibuya1 and Hiroshi Kodama2 1 JGC Corporation, 2205, Naritacho, Oaraimachi, 311-1313, Japan 2 National Institute for Research in Inorganic Materials, Namiki 1-1, Tsukuba, Ibaraki 305-0044, Japan
ABSTRACT Leaching of the iodide ion from BiPbO2I (BPI), BPI encapsulated in cement (BPIC) and AgI was studied in a low salinity solution and in a high salinity solution under reducing conditions. Although BPI released a limited amount of iodide ions (less than 1%) into the low salinity solution, it released more than 30% of iodide ions into the high salinity solution within 80 days. AgI released more than 30% of iodide ions into both low and high salinity solutions within 80 days. It was proved that BPI is more stable than AgI in the low salinity solution under reducing conditions. BPIC released a limited number of iodide ions (less than 5%) into both low and high salinity solutions. BPIC showed the best leach resistance in the high salinity solution. BiPbO2NO3 (BPN) was developed to remove iodide ions in a solution and fix them in BPI by the ion exchange reaction. Ion exchange properties under reducing conditions were studied. An anion exchange capacity of 1.0 mEq/g and a distribution coefficient of larger than 0.1 m3/kg were obtained in a solution at a pH of between 9 and 13. The advantages of the process using BPN for removing and immobilizing Iodine-129 were discussed from the standpoint of process simplification.
INTRODUCTION Radioactive wastes containing Iodine-129 are to be disposed of in an underground facility together with TRU wastes, in Japan. Iodine-129 has a long half-life (1.6× 107 y) and it strongly adsorbed in the thyroid gland when it intrudes into the human body [1]. The main chemical forms of iodine in an alkaline solution are I- and IO3-, and these anion species are rarely absorbed on silicate minerals. Therefore, Iodine-129 is one of the key nuclides to be studied in the geological disposal of radioactive wastes [2,3]. A preliminary safety assessment of a TRU waste disposal system was reported [4] and it showed that the dose rate of Iodine-129 was far larger than that of other nuclides when iodine was not sufficiently immobilized. Therefore, it is important to develop technologies to treat and dispose of waste containing Iodine-129 from the view point of achieving an enhanced level of confidence in radioactive waste management. L. L. Burger, R. D. Scheele et al. surveyed thermodynamic data on both iodide and iodate compounds. They recommended AgI, Ba(IO3)2 and Ca(IO3)2 encapsulated in a cement matrix as candidate wasteforms [5-6]. D. L. Brown, M. W. Grutzeck et al. recommended immobilizing Iodine-129 into tetracalcium aluminate monosulfate–12–hydrate (3CaOAl2O3Ca(IO3)2 !12H2O).
They reported that the leach rate is near or close to that of Ba(IO3)2 encapsulated in cement [7]. E. R. Vance et al. studied iodine-sodalite (Na8Al6Si6O24I2) [8]. Inorganic materials of iodine getter were also investiga
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