• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.266-269
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• 2003

To understand the long-term leach behavior of a borosilicate waste glass in a repository, the leaching experiment with three kinds of simulated borosilicate waste glasses has been carried out since the middle of 1997. The five years results indicate that a boron would be applied as an indicator of a long-term leaching of their borosilicate waste glasses and that their long-term leach rates have a tendency to be close to about $0.03g/m^2-day$ even though their compositions and their ratios of the surface area to the volume of leachate are different.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.41-46
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• 2003

To understand the long-term leach behavior of a borosilicate Waste glass in a repository, the leaching experiment with three kinds of simulated borosilicate waste glasses has been carried out since the middle of 1997. The five years results indicate that a boron would be applied as an indicator of a long-term leaching of their borosilicate waste glasses and that their long-term leach rates have a tendency to be close to about 0.03g/$m^2$-day even though their compositions and their ratios of the surface area to the volume of leachate are different.

• Journal of the Korea institute for structural maintenance and inspection
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• v.22 no.2
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• pp.16-25
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• 2018

Concrete is a suitable construction material for long-term structure, however, it is needed to understand the calcium leaching damage caused by exposure to underground pure water for a long time. In this paper, it is experimentally investigated that the characteristics of flexural behavior in RC member damaged by calcium leaching degradation. From the test results, when calcium leaching is happened, yielding load and flexural rigidity is reduced, neutral axis depth and displacement is increased. That is, calcium leaching degradation adversely affects RC member performance. And, when the mineral admixture is used in the calcium leaching environment, it is considered that the optimal replacement ratio should be prepared according to the type of mineral admixture.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.197-202
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• 1988

A finite slab approximation method was developed to predict the long-term teachability. It is based on the assumption that the diffusional characteristics of radionuclides in a waste matrix are not dependent on matrix geometry but dependent on volume to surface ratio V/S) and diffusion coefficient. Consequently it can be expressed as the solution of the equations obtained from a finite slab with an equal V/S ratio (imaginary diffusion length). The calculational results by the finite slab approximation method have been compared with the results obtained for finite cylinder and sphere with corresponding diffusional analysis. The results of this simple model have showed a good agreement and presented a general applicability for the long-term prediction of the radionuclide leaching behavior.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.3
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• pp.323-330
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• 2021

Silver tellurite glasses with melting temperature of approximately 700℃ were developed to immobilize ¹²⁹I wastes. Long-term dissolution tests in 0.1 M acetic acid and disposability assessment were conducted to evaluate sustainability of the glasses. Leaching rate of Te, Bi and I from the glasses decreased for up to 16 d, then remained stable afterwards. On the contrary, tens to tens of thousands of times more of Ag was leached in comparison to the other elements; additionally, Ag leached continuously for all 128 d of the test owing to the exchange of Ag⁺ and H⁺ ions between the glasses and solution. The I leached much lower than those of other elements even though it leached ~10 times more in 0.1 M acetic acid than in deionized water. Some TeO₄ units in the glass network were transformed to TeO₃ by ion exchange and hydrolysis. These silver tellurite glasses met all waste acceptance criteria for disposal in Korea.

• Nuclear Engineering and Technology
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• v.28 no.4
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• pp.349-358
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• 1996

The leaching behavior of uranium from unirradiated CANDU UO$_2$ fuel pellet in the spent fuel wet storage and disposal conditions has been investigated. A modified IAEA leach test method was used, and then the extent of leaching was monitored by analysis for uranium in the leachant. The leach test has been performed in various leachants(demineralized water and boric acid solution at pH=6, synthetic granite groundwater) for a long-term period of 5.4 years, and the effect of temperature on the leach rate of uranium has been analyzed. The leach rates of uranium at $25^{\circ}C$ were dependent on the leachants. Over initial 100 days of leach periods, the leach rate in groundwater was the highest in three leachants and no significant differences of leach rates ore observed in the demineralized oater and boric acid solution. But these leach rates in three leachants around 2,000 days at $25^{\circ}C$ appeared to be reached the steady rates in the range of 1~5$\times$10$^{-8}$ g/$\textrm{cm}^2$ day. The leach rate of uranium in groundwater shooed to be independent of the temperature, but those in both demineralized water and boric acid solution increased with temperature. These results show that the leaching behavior of uranium from UO$_2$ fuel in both the demineralized water ann boric acid may be controlled tv the surface oxidative.dissolution reaction of UO$_2$ and the leach rate of uranium in groundwater at room temperature could mainly be controlled by the complex reaction of dissolved uranyl ions with carbonate ions and no variation of leach rate of UO$_2$ in groundwater with temperature may be due to the local deposition of passivating uranyl phases on the surface.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.167-173
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• 2004

The amount of cesium released from the leaching of spent fuels in contact with and without the compacted bentonite bloc]t which was compacted as the density of $1.4g/\textrm{cm}^3$, up to 5.7 years were measured and the empirical formula of the fractional release rate of cesium were derived from these measured values. The empirical formulas show that the long-term release rate of cesium under a repository would become a constant, as about $3{\times}10_{-6}$ fraction/day, after a certain period. The cumulative fractions of cesium released from the spent fuel with bentonite and with copper and stainless steel sheets were steadily increased, but the fraction from bare fuel was rapidly increased and then sluggishly increased. However, the remained value except its gap inventory from the cumulative fraction of cesium released from bare fuel was almost very close to the others. This suggests that the initial release of cesium from bare fuel might be dependant on its gap inventory.

• Journal of Radiation Protection and Research
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• v.11 no.2
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• pp.91-103
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• 1986

Experimental study for the leaching behavior of Cs-137 was carried out using the simulated evaporator bottom product of PWR plant. The method of leach test proposed by the IAEA was partially modified using ANS method. The effect of various factors, i.e., sampling method, curing temperature, curing time, leachant temperature, vermiculite addition and volume-to-surface ratio, was considered in this experiment. Diffusion model in semi-infinite slab was in a good agreement with the data obtained from 4-weeks cured specimens. The effective diffusion coefficient of the specimens which were cured at the temperature of $24^{\circ}C$ for 4 weeks was found to be $1.20{\sim}1.47{\times}10^{-11}cm^2/sec$. With the experimentally obtained diffusion coefficient ($1.47{\times}10^{-11}cm^2/sec$), long-term prediction for the leaching of Cs-137 was carried out using finite-slab approximation. The estimated fraction of Cs-137 which remains in the environment is found to be less than 0.25 percent of initial amount after 100 years. About 25 years after the beginning of leaching, its fractional amount in the environment reachs the maximum value, 0.66 percent of initial amount.