Effects of Bi2O3, Al2O3, PbO on silver tellurite glass for radioactive iodine immobilization

Kang, Hyun Woo; Lee, Ki Rak; Choi, Jonghyun; Park, Hwan-Seo

doi:10.1007/s10967-020-07421-0

Cited by 7 publications

(2 citation statements)

References 14 publications

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“…As in the case of other glass wasteforms, the use of crosslinking agents such as Bi 2 O 3 , Al 2 O 3 , and PbO has been investigated to try and reduce the breakdown of the tellurite glass network observed during leach studies (Kang et al, 2020). When the mass fraction of Bi 2 O 3 and PbO are 10%, an order of magnitude reduction in Te release rates was observed.…”

Section: Tellurite Glassmentioning

confidence: 99%

Review of recent developments in iodine wasteform production

et al. 2022

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Radioiodine capture and immobilization is not only important to consider during the operation of reactors (i.e., I-131), during nuclear accidents (i.e., I-131 and I-129) or nuclear fuel reprocessing (i.e., I-131 and I-129), but also during disposal of nuclear wastes (i.e., I-129). Most disposal plans for I-129-containing waste forms (including spent nuclear fuel) propose to store them in underground repositories. Here, iodine can be highly mobile and, given its radiotoxicity, needs to be carefully managed to minimize long-term environmental impacts arising from disposal. Typically, any process that has been used to capture iodine from reprocessing or in a reactor is not suitable for direct disposal, rather conversion into a wasteform for disposal is required. The objectives of these materials are to use either chemical immobilization or physical encapsulation to reduce the leaching of iodine by groundwaters. Some of the more recent ideas have been to design capture materials that better align with disposal concepts, making the industrial processing requirements easier. Research on iodine capture materials and wasteforms has been extensive. This review will act as both an update on the state of the research since the last time it was comprehensively summarized, and an evaluation of the industrial techniques required to create the proposed iodine wasteforms in terms of resulting material chemistry and applicability.

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Section: Tellurite Glassmentioning

confidence: 99%

Review of recent developments in iodine wasteform production

et al. 2022

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“…Recently, this waste form has been actively researched to find ways to solidify the I nuclide. Notably, Ag 2 O-AgI-M 2 O 3 , where M is a metal cation, is considered because the glass system can accommodate I nuclides with high waste loading compared to the iodoapatite ceramic waste form [19][20][21]. Since the composition could be diversified from the glass system, the waste loading was not fixed.…”

Section: Characterization Of I Waste/wastementioning

confidence: 99%

Characterization of Waste Generated from Nuclide Management Process in Waste Burden Minimization Technology for Spent Nuclear Fuel

Choi

Lee

et al. 2022

Science and Technology of Nuclear Installations

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To reduce the environmental burden caused by the disposal of spent nuclear fuel, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. To manufacture a waste form of high durability, the characteristics of the waste generated during the process should be evaluated. In this study, the physical, radiological, and thermal characteristics of the waste and waste forms for major nuclides (Cs, Sr, I, transuranic/rare earth, and Tc/Se) generated in the nuclide management process were analyzed. In the case of Cs nuclides, characterization was conducted according to the capture rate of the adsorbent in the high-temperature heat treatment process; meanwhile, in the case of Sr nuclides, characterization was performed by considering the ratio of similar nuclides in the chlorination process. For I nuclide, analysis was performed based on the available waste form, and for TRU/RE and Tc/Se nuclides, analysis was performed by considering chlorination and mid-temperature heat treatment. The radioactivity and heat generation rate of each waste and waste form were evaluated over a period of 1,000 years. The results of this study could be used to derive the centerline temperature for the thermal stability evaluation of waste forms and for the feasibility evaluation of each disposal system considered in the waste burden minimization technology.

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