• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.228-235
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• 1992

Some soluble species may not be solubility-limited or congruent-released with the matrix species. For example, during the operation of the nuclear reactor, the fission products can be accumulated in the fuel-cladding gap, voids, and grain boundaries of the fuel rods. In the waste package for spent-fuel placed in a geologic repository, the high solubility species of these fission products accumulated in the“gap”, e.g. cesium or iodine are expected to dissolve rapidly when ground water penetrates fuel rods. The time and space dependent mass transport for high solubility nuclides in the gap is analyzed, and its numerical illustrations are demonstrated. The approximate solution that is valid for all times is developed, and validated by comparison with an asymptotic solution and the solution obtained by the numerical inversion of Laplace transform covering the entire time span.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.302-307
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• 1989

Nuclide release through penetrations in radioactive waste container is analyzed. Penetrations may result from corrosion or cracking and may be through the container material or through deposits of corrosion products. The analysis deals with the resultant nuclide release, but not with the way these penetrations occur. Numerical illustrations show that mass transport from multiple holes can be significant and may approach the mass transfer rate calculated from bare waste forms. Although partially-failed containers may present an important long-term barrier to release of radionuclides, numerous small holes on a container surface have the potential of bypassing the effectiveness of these barriers.

• Journal of Radiation Protection and Research
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• v.21 no.1
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• pp.51-58
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• 1996

Underground migration of $^{54}Mn,\;^{60}Co,\;^{85}Sr\;and\;^{137}Cs$ in paddy and upland conditions was studied through two years' greenhouse experiment. At early and late growth stages of rice, soybean, Chinese cabbage and radish, a mixed solution of the radionuclides was applied to the water or soil surfaces of the culture boxes filled with an acidic loamy-sandy soil for the upper 20cm. Soil was sampled in layers upto $15{\sim}20cm$ down after harvest. Soil concentrations of the radionuclides decreased exponentially with increasing soil depth and more than 80% of the radioactivities remained in top $3{\sim}4cm$. The mobility of the radionuclides decreased in the order of $^{85}Sr>^{54}Mn>^{60}Co{\geq}^{137}Cs$. Downward migrations of the radionuclides were the greatest in rice soil and the lowest in soybean soil which was fertilized with the least amount of N, P and K. Differences in depth profiles between two application times indicate that the amount of daily migration from $0{\sim}1cm$ layer to the lower area decreases with increasing time after deposition. By a simultaneous addition of KCl and lime following the earlier application, downward migration in soybean, Chinese cabbage and radish soils changed little or retarded more or less but that in rice soil accelerated a little.

• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.387-400
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• 1991

An ocean compartment model simulating transport of nuclides by advection due to ocean circulation and intertaction with suspended sediments is developed, by which concentration breakthrough curves of nuclides can be calculated as a function of time. Dividing ocean into arbitrary number of characteristic compartments and performing a balance of mass of nuclides in each ocean compartment, the governing equation for the concentration in the ocean is obtained and a solution by the numerical integration is obtained. The integration method is specially useful for general stiff systems. For transfer coefficients describing advective transport between adjacent compartments by ocean circulation, the ocean turnover time is calculated by a two-dimensional numerical ocean model. To exemplify the compartment model, a reference case calculation for breakthrough curves of three nuclides in low-level radioactive wastes, Tc-99, Cs-137, and Pu-238 released from hypothetical repository under the seabed is carried out with five ocean compartments. Sensitivity analysis studies for some parameters to the concentration breakthrough curves are also made, which indicates that parameters such as ocean turnover time and ocean water volume of compartments have an important effect on the breakthrough curves.

• Journal of Radiation Protection and Research
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• v.16 no.2
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• pp.27-39
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• 1991

The diffusion of Sr-85, Cs-137, Co-60 and Am-241 in compacted domestic bentonite was studied, using a diffusion cell unit in which diffusion took place axially from the center of cylindrical bentonite sample body. The effects of compaction density and heat-treated bentonite on diffusion were analysed. And the diffusion mechanism of radionuclide was also analysed by evaluating the measured diffusivity of anion Cl-36. The apparent diffusivities obtained for Sr-85, Cs-137, Co-60 and Am-241 were $l.07{\times}10^{-11},\;6.705{\times}10^{-13},\;l.226{\times}10^{-13}\;and\; l.310{\times}10^{-14}m^2/sec$, respectively. When the as-pressed density of bentonite increased from $1.8\;to\;2.0g/cm^3$, the apparent diffusivity of Cs-137 decreased by quarter. In the case of bentonite heat-treated to $150^{\circ}C$, no significant change in diffusivity was observed, which showed the possibility that the domestic bentonite could be used as a chemical barrier to retard the radionuclide migration at below $150^{\circ}C$. From the calculated pore and surface diffusivity, the surface diffusion due to the concentration gradient of radionuclide sorbed on the solid phase was found to dominate greatly in total transport process.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.539-547
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• 1993

A stochastic approach using continuous time Markov process is presented to model the one-dimensional nuclide transport in fractured rock media as a further extension for previous works［1-3］. Nuclide transport of decay chain of arbitrary length in the single planar fractured rock media in the vicinity of the radioactive waste repository is modeled using a continuous time Markov process. While most of analytical solutions for nuclide transport of decay chain deal with the limited length of decay chain, do not consider the case of having rock matrix diffusion, and have very complicated solution form, the present model offers rather a simplified solution in the form of expectance and its variance resulted from a stochastic modeling. As another deterministic way, even numerical models of decay chain transport, in most cases, show very complicated procedure to get the solution and large discrepancy for the exact solution as opposed to the stochastic model developed in this study. To demonstrate the use of the present model and to verify the model by comparing with the deterministic model, a specific illustration was made for the transport of a chain of three member in single fractured rock medium with constant groundwater flow rate in the fracture, which ignores the rock matrix diffusion and shows good capability to model the fractured media around the repository.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.41-48
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• 1991

Many of the analyses of the transient radionuclide migration are approximated by an one-dimensional geometry and/or planar geometry. To validate these approximations, one should prove that these are reasonable and proper approximations. In this paper, the approximation which was in the study of the transport through backfill into a fissure is tried to validate. In that analysis, a cylindrical geometry was approximated by a planar geometry. The numerical illustrations show that the planar approximation agrees very well with the result of the cylindrical geometry for a ratio of the backfill outer radius to the waste form radius closed to unity. Even for a larger ratio of the two radii, the numerical difference is relatively small. Also the planar approximation which was used in the analysis gives conservative estimates.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.582-590
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• 1995

우리 나라에서 건설될 중저준위폐기물 처분장의 기본설계개념에 따르면, 중준위폐기물 동굴에는 점토질 뒷채움재가 설치될 예정이다. 이 뒷채움재의 후보물질로는 원료 입수의 용이성 및 경제성 측면에서 천연점토와 분쇄암석의 혼합물이 고려되고 있다. 그러나 이 혼합물이 폐기물처분장의 뒷채움재로 사용되기 위해서는 주요특성들이 뒷채움재의 요구사항을 만족시켜야한다. 이 보고서에서는 경상북도 연일지역에서 산출되는 천연점토와 분쇄암석 혼합물에 대해 핵종유출저지 관점에서 가장 중요한 특성인 수리특성과 방사성핵종 확산특성 및 역학적 특성을 측정하여 뒷채움재료로서의 사용 가능성을 평가하였다. 혼합물의 수리전도도는 동일 밀도에서 점토함량이 감소함에 따라 증가되었으나, 점토함량이 적은 경우에도 상당히 낮은 값을 유지하여 뒷채움재를 통한 핵종이동의 주요 메커니즘은 확산이 될 것이라는 것을 보여주었다. 압축점토에 대한 방사성핵종 복산계수 측정결과, 동일밀도의 나트륨벤토나이트에서 측정된 값의 범위 내에 있었으며, 점토-분쇄암석 혼합물의 액소성한계, 다짐특성, 압축강도및 압밀계수등의 역학적 특성도 나트륨벤토나이트-모래 혼합물에서 얻어진 값과 유사하였다. 이러한 결과들로부터 연일산 천연점토-분채암석 혼합물을 고가의 나트륨벤토나이트 대용으로 처분장 뒷채움재로 사용할 수 있다는 것을 알 수 있었다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.99-105
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• 2011

An integrated model for groundwater flow and radionuclide transport analyses is being developed incorporating six underground silos, an excavated damaged zone (EDZ), and fractured host rock. The model considers each silo as an engineered barrier system (EBS) consisting of a waste zone comprising waste packages and disposal container, a buffer zone, and a concrete lining zone. The EDZ is the disturbed zone adjacent to silos and construction & operation tunnels. The heterogeneity of the fractured rock is represented by a heterogeneous flow field, evaluated from discrete fractures in the fractured host rock. Radionuclide migration through the EBS in silos and the fractured host rock is simulated on the established heterogeneous flow field. The current model enables the optimization of silo design and the quantification of the safety margin in terms of radionuclide release.

• Journal of Radiation Protection and Research
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• v.26 no.4
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• pp.375-384
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• 2001

It is possible to count and perform quench correction on two ${\beta}$ -label samples so long as the maximum ${\beta}$-energies are sufficiently different. However, when 4he conventional technique is applied to the radioassay of a mixture of more than three nuclides, the reliability of the activities determined is considerably reduced, resulting from the large overlapping of liquid scintillation pulse height distributions of each nuclide. A technique that allows the activities of multiple ${\beta}$-labeled samples to be radioassayed was proposed by using the least square method. The technique was applied to mixture samples of $^3H,\;^{14}C,\;^{36}Cl$, and $^{90}Sr$. The analytical values were in good agreement with the reference values within 7% relative error.