• 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.40 no.7
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• pp.593-606
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• 2008

In this study, data on radionuclide migration and retardation processes in the engineered and natural barriers of High-Level Radioactive Waste (HLW) repository have been reviewed and compiled for use in the performance assessment of a HLW disposal system in Korea. The status of the database on radionuclide migration and retardation that is being developed in Korea is investigated and summarized in this study. The solubilities of major actinides such as D, Th, Am, Np, and Pu both in Korean bentonite porewater and in deep Korean groundwater are calculated by using the geochemical code PHREEQC (Ver. 2.0) based on the KAERI-TDB(Korea Atomic Energy Research Institute-Thermochemical Database), which is under development. Databases for the diffusion coefficients ($D^b_e$ values) and distribution coefficients ($K^b_d$ values) of some radionuclides in the compacted Korean Ca-bentonite are developed based upon domestic experimental results. Databases for the rock matrix diffusion coefficients ($D^r_e$ values) and distribution coefficients ($K^r_d$ values) of some radionuclides for Korean granite rock and deep groundwater are also developed based upon domestic experimental results. Finally, data related to colloids such as the characteristics of natural groundwater colloids and the pseudo-colloid formation constants ($K_{pc}$ values) are provided for the consideration of colloid effects in the performance assessment.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.276.1-276
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• 2001

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.819-826
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• 2020

The method of the Laplace transform has been used to obtain an analytical solution of the three-dimensional steady state advection diffusion equation for the airborne radionuclide release from any nuclear installation such as the power reactor in an area source. The present treatment takes into account the removal of the pollutants through the nuclear reaction. We assume that the pollutants are emitted as a constant rate from the area source. This physical consideration is achieved by assuming that the vertical eddy diffusivity coefficient should be a constant. The prevailing wind speed is a constant in 𝑥- direction and a linear function of the vertical height z. The present model calculations are compared with the other models and the available data of the atmospheric dispersion experiments that were carried out in the nuclear power plant of Angra dos Reis (Brazil). The results show that the present treatment performs well as the analytical dispersion model and there is a good agreement between the values computed by our model and the observed data.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1449-1455
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• 2022

A radiotracer investigation was carried out in a Thermal Stratification Test Facility (TSTF) with objectives of investigating the dispersion and diffusion of radionuclide and effectiveness of the thermocline to minimize the radionuclide content in the hot water layer. Technetium-99m (^99mTc) as sodium pertechnetate was used as a radiotracer in the investigation. Qualitative analysis showed that a thermocline is formed within the TSTF and is effective in preventing the transport of radionuclide from bottom section to the top section of the facility. It was found that the radiotracer injected at the bottom of the pool took about 17.4 h to disperse from bottom to the top of the facility. The results of the investigation helped in understanding the effectiveness of hot water layer and thus to minimize the pool top radiation levels.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.274-279
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• 1996

Diffusion coefficient is a critical parameter for predicting radiological source term(migration rate and flux of radionuclide) through given near field conditions in spent fuel or high level waste repository. The effect of exchangeable cation-$Na^+$ and $Ca^{2+} - on the diffusion of $I^- \;and^3H$ (as HTO) in compacted bentonite was examined using a through-diffusion method. Bentonite material used here was compacted to a density of 1.3 Mg/m$^3$, and Na-bentonite was saturated with a solution of 100 mol NaCl/m$^3$ and Ca-bentonite with 50 $mol\;CaCl_2$/m$^3$. The results show that effective diffusion coefficients are generally higher by a factor of two to five in Ca-than Na-clay. This is attributed to the larger particle size of Ca-compared to Na-bentonite; hence, Ca-bentonite has a greater proportion of relatively large pores, which make a greater contribution to mass transport than small pores. Although the nature of the exchangeable cation affects mass diffusion in compacted bentonite, the effect is small and not likely to influence performance assessment modeling of compacted bentonite-based barriers.

• Journal of Radiation Protection and Research
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• v.29 no.1
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• pp.41-48
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• 2004

A safety assessment code, called SAGE (Safety Assessment Groundwater Evaluation), has been developed to describe post-closure radionuclide releases and potential radiological doses for low- and intermediate-level radioactive waste (LILW) disposal in an engineered vault facility in Korea. The conceptual model implemented in the code is focused on the release of radionuclide from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. The radionuclide transport equations are solved by spatially discretizing the disposal system into a series of compartments. Mass transfer between compartments is by diffusion/dispersion and advection. In all compartments, radionuclides ate decayed either as a single-member chain or as multi-member chains. The biosphere is represented as a set of steady-state, radionuclide-specific pathway dose conversion factors that are multiplied by the appropriate release rate from the far field for each pathway. The code has the capability to treat input parameters either deterministically or probabilistically. Parameter input is achieved through a user-friendly Graphical User Interface. An application is presented, which is compared against safety assessment results from the other computer codes, to benchmark the reliability of system-level conceptual modeling of the code.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.361.2-361
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.344.3-344
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• 2002

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.237-246
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• 2012

In this study, numerical model for transport of radionuclide and colloid was developed. In order to solve reaction-migration governing equation for colloid and radionuclide, Strang-splitting Sequential Non-Iterative (SNI), which is one of Operator Splitting Method, was used for numerical method and this was coded by MATLAB. From the verification by comparing the simulation results with analytical solution considering only solute transport and rock diffusion, the Pearson's correlation coefficient was greater than 0.99 which demonstrates the accuracy of the model.