• Nuclear Engineering and Technology
• /
• v.34 no.3
• /
• pp.218-231
• /
• 2002

Using a three-dimensional numerical code, B3R developed for nuclide transport of an arbitrary length of decay chain in the buffer between the canister and adjacent rock in a high- level radioactive waste repository by adopting a finite difference method utilizing the control- volume scheme, some illustrative calculations have been done. A linear sorption isotherm, nuclide transport due to diffusion in the buffer and the rock matrix, and advection and dispersion along thin rigid parallel fractures existing in a saturated porous rock matrix as well as diffusion through the fracture wall into the matrix is assumed. In such kind of repository, buffer and rock matrix are known to be important physico-chemical harriers in nuclide retardation. To show effects of buffer and rock matrix on nuclide transport in HLW repository and also to demonstrate usefulness of B3R, several cases of breakthrough curves as well as three- dimensional plots of concentration isopleths associated with these two barriers are introduced for a typical case of decay chain of $^{234}$ Ulongrightarrow$^{230}$ Thlongrightarrow$^{226}$ Ra, which is the most important chain as far as the human environment is concerned.

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

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.05a
• /
• pp.225-226
• /
• 2017

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

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

• Tunnel and Underground Space
• /
• v.13 no.1
• /
• pp.20-32
• /
• 2003

To understand the behavior of migration of contaminants in a fractured porous medium is a key to assure the overall safety of a potential radwaste repository. The feasible retention mechanism of contaminant transport in a tinctured medium are sorption of contaminants on solid surface and matrix diffusion of contaminants from a fracture into an adjacent porous medium. The acceleration mechanisms are the migration of contaminants in the form of pseudo-colloids and the limit of a volume f3r matrix diffusion. In this paper, the effects of these two acceleration mechanisms are studied mathematically, then semi-analytically computed by the application of the Talbot theorem and verified. Results indicate that the acceleration processes cannot be neglected in the modeling of contaminant transport in a fractured porous medium.

• Nuclear Engineering and Technology
• /
• v.40 no.7
• /
• pp.593-606
• /
• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.4
• /
• pp.281-291
• /
• 2013

A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.

• The Pure and Applied Mathematics
• /
• v.22 no.2
• /
• pp.159-168
• /
• 2015

Abstract. We propose a fast and robust finite difference method for Merton's jump diffusion model, which is a partial integro-differential equation. To speed up a computational time, we compute a matrix so that we can calculate the non-local integral term fast by a simple matrix-vector operation. Also, we use non-uniform grids to increase efficiency. We present numerical experiments such as evaluation of the option prices and Greeks to demonstrate a performance of the proposed numerical method. The computational results are in good agreements with the exact solutions of the jump-diffusion model.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.255-263
• /
• 2005

Migration experiments of THO and 237Np have performed through a sampled granite core in Chemlab2 probe at the Aspo hard Rock laboratory. The elution curves of THO were analysed to determine hydraulic properties such as the extent of dispersion effect according to flow rates. The retardation phenomena of the solutes were observed and described with elution curves and migration plumes. After migration test, the rock core was opened, and the remaining radioactivities on the rock fracture surfaces were measured. The transport process was simulated with a two-dimensional channel model. The mass transport process was described with three types of basic processes ; advection, sorption and matrix diffusion. By the combination of these processes, effects of each process on transport were described in terms of elution curves and migration plumes. By comparing the simulation results to the experimental one, it was possible to analyse the retardation effect quantitatively.