The n-type Bi2-xSbxTe3 compounds have been of great interest due to its potential to achieve a high thermoelectric performance, comparable to that of p-type Bi2-xSbxTe3. However, a comprehensive understanding on the thermoelectric properties remains lacking. Here, we investigate the thermoelectric transport properties and band characteristics of n-type Bi2-xSbxTe3 (x = 0.1 – 1.1) based on experimental and theoretical considerations. We find that the higher power factor at lower Sb content results from the optimized balance between the density of state effective mass and nondegenerate mobility. Additionally, a higher carrier concentration at lower x suppresses bipolar conduction, thereby reducing thermal conductivity at elevated temperatures. Consequently, the highest zT of ~ 0.5 is observed at 450 K for x = 0.1 and, according to the single parabolic band model, it could be further improved by ~70 % through carrier concentration tuning.

The effect of various physicochemical processes, such as seawater intrusion, on the performance of the engineered barrier should be closely analyzed to precisely assess the safety of high-level radioactive waste repository. In order to evaluate the impact of such processes on the performance of the engineered barrier, a thermal-hydrological-chemical model was developed by using COMSOL Multiphysics and PHREEQC. The coupling of two software was achieved through the application of a sequential non-iterative approach. Model verification was executed through a comparative analysis between the outcomes derived from the developed model and those obtained in prior investigations. Two data were in a good agreement, demonstrating the model is capable of simulating aqueous speciation, adsorption, precipitation, and dissolution. Using the developed model, the geochemical evolution of bentonite buffer under a general condition was simulated as a base case. The model domain consists of 0.5 m of bentonite and 49.5 m of granite. The uraninite (UO2) was assigned at the canister-bentonite interface as the potential source of uranium. Assuming the lifetime of canister as 1,000 years, the porewater mixing without uranium leakage was simulated for 1,000 years. After then, the uranium leakage through the dissolution of uraninite was initiated and simulated for additional 1,000 years. In the base case model, where the porewater mixing between the bentonite and granite was the only considered process, the gypsum tended to dissolve throughout the bentonite, while it precipitated in the vicinity of bentonite-granite boundary. However, the precipitation and dissolution of gypsum only showed a limited effect on the performance of the bentonite. Due to the low solubility of uraninite in the reduced environment, only infinitesimal amounts of uranium dissolved and transported through the bentonite. Additional cases considering various environmental processes, such as seawater or cement porewater intrusion, will be further investigated.

Given the situation in the Republic of Korea that all nuclear power plants are located at the seaside, the interim storage facility is also likely to be located at seaside and the maritime transportation of Spent Nuclear Fuel is considered inevitable. The Republic of Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radionuclides from a submerged transportation cask in the sea. Therefore, there is a need to develop a technology that can assess the impact of immersion accidents and establish a regulatory framework for maritime transportation accidents. The release rate of radionuclides should be calculated from the flow rate through a flow path in the breached containment boundary. According to the cask design criteria, it is anticipated that even under severe accident conditions, the flow path size will be very small. Previous studies have evaluated fluid flow passing through micro-scale channel by integrating internal and external flows within and around a transport cask. As part of the evaluation, a comprehensive “Full-Field Model” incorporating external flow fields and a localized “Local-Field Model” with micro-scale flow paths were constructed. Sub-modeling techniques were employed to couple the flow field calculated by the two models. The aforementioned approach is utilized to conduct the evaluation of fluid flow passing through micro-scale flow paths. This study aims to evaluate fluid flow passing through micro-scale flow paths using the aforementioned CFD (Computational Fluid Dynamics) method and aims to code the findings. The Gaussian Process Regression technique, a machine learning model, is utilized for developing a mathematical metamodel. The selected input parameters for coding are organized and their respective impacts are analyzed. The range of these selected parameters is tailored to suit domestic environments, and computational experiments are planned through Design of Experiments. The flow path size is included as an input parameter in the coded model. In cases where the flow path size becomes extremely small, making it impractical to use CFD techniques for calculations, Poiseuille’s law is employed to calculate the release rate. In this study, a model is developed to evaluate the release rate of radionuclides using CFD and mathematical equations covering the whole possible range of flow path size in a lost cask in the deep sea. The model will be used in the development of a maritime transportation risk assessment code suitable for the situation and environment in Korea.

LILW disposal repository in Gyeongju, South Korea is considered with a concrete mixture that uses Ordinary Portland Cement (OPC) partially substituted with supplementary cementitious materials (SCMs). The degradation of cementitious materials that result from chemical and physical attacks is a major concern in the safety of radioactive waste disposal. We present a reactive transport model utilized as one of the geochemical simulation approaches for the timescales of concern that range from hundreds to thousands of years. The purpose of this study is to investigate the sensitivity of parameters in concrete disposal systems and to evaluate the influence of various assumptions on the chemical degradation of the systems using a reactive transport model. A reactive transport model in the concrete disposal vault was developed to evaluate the behavior of engineered barriers composed of cementitious materials. The sensitivity analysis was performed using reactive transport models through the coupling between COMSOL and PHREEQC. The databases selected for the analysis are the Thermochimie database presented by ANDRA. Among many variables considered, two variables that can highly affect chemical degradation were selected for detailed sensitivity analysis for dealing with uncertainties. This is important because the chemical degradation mechanism is generally sensitive to precipitation and diffusion coefficient. The first factor is precipitation, which might be the most important factor in chemical degradation because it acts as a calcium leaching of cementitious materials in a disposal system in a highly alkaline environment, increasing the porosity of the system. To predict the change in annual precipitation, the measurement of the precipitation observatory station in the nearest area of Gyeongju for the past 80 years was collected. The second factor is the diffusion coefficient, which plays an essential role in the durability of the concrete disposal system, promoting the decalcification of cementitious minerals, accelerating system degradation, and increasing the porosity of its system, thereby facilitating the migration of radionuclides. The diffusion coefficient values used in studies similar to this work were calculated and evaluated using the box-and-whisker method. The results of the sensitivity analyses for the reactive transport model in the concrete disposal system will be presented. The sensitivity cases show that the results obtained are much more sensitive to changes in transport parameters.

According to the ‘Basic Plan for High-Level Radioactive Waste Management (draft)’, the total amount of CANDU spent nuclear fuel is expected to be approximately 660,000 bundles. To safely and efficiently transport this amount to interim storage facilities, it is essential to develop a large-capacity transport cask. Therefore, we have been developing a large-capacity PHWR spent nuclear fuel transport cask, called the KTC-360 transport cask. According to the transport-cask related regulations, the KTC-360 transport cask was classified as a Type B package, and such packages must be able to withstand a temperature of 800°C for a period of 30 min. It is desirable to conduct a test using a fullscale model of a shipping package when performing tests to evaluate its integrity. However, it is costly to perform a test using a full-scale model. Therefore, to evaluate the thermal integrity of the KTC-360 transport cask, the fire test was conducted using a slice model. For comparison purposes, the fire test was also carried out using a 1/4 scale model. In the fire test using a slice model and in the fire test using a 1/4 scale model, the maximum temperature of the cask body was lower than the permitted maximum temperature limit. Therefore, the thermal integrity of the KTC-360 transport cask could be considered to be maintained. The temperature results from the fire test using a slice model were higher than those of the fire test using a 1/4 scale model. Therefore, the effect of flame on a transport cask without combustible materials, such as the KTC-360 transport cask, seems to be affected by the reduction in the time rather than the size reduction.

As the saturation rate of temporary storage facilities for spent nuclear fuel increases, regulatory demands such as interim storage and permanent disposal of spent nuclear fuel are expected to begin in earnest. Considering the domestic situation where all nuclear power plants are located on the waterfront site, the interim storage site is also likely to be located on the waterfront site, and maritime transportation is one of the essential management stages. Currently, there are no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the sinking of transport container. Therefore, it is necessary to secure technology to properly reflect the domestic maritime transportation environment and to assess the impact of the sinking accident and to carry out safety regulations. To accurately calculate the releaser rate of radionuclides contained in a cask with breached containment boundary, the flow rate through the gap generated in the containment boundary should be calculated. The fluid flow through this gap which is probably in micro scale in most situations should be evaluated combining the fluid flow inside and outside the cask. In this study, a detailed computational fluid dynamics model to evaluate the internal fluid flow in the cask and a simplified model to capture the fluid flow and the heat transfer around the cask in the sea are constructed. The results for the large scale model are compared with the analytic formula for verification of heat transfer coefficient and they showed good agreements. The heat transfer coefficient thus found can be used in the detailed model to provide more realistic data than those obtained from assumed heat transfer coefficient around the surface of the cask. In the future, fluid flow through the gap between the lid and the body of the cask will be evaluated coupling the models developed in this work.

An elliptic blending Reynolds stress transport equation model for Newtonian fluids has been extended to predict polymer-induced drag reduction FENE-P fluids. The conformation tensor equation which is related to the polymer stress is adopted from the model form of Resende et al., and the models of redistribution and dissipation rate terms for the Reynolds stress transport equation are considered by the elliptic blending equation. Also, the new model terms for viscoelastic turbulent transport and viscoelastic dissipation in the Reynolds stress transport equation are introduced to consider the polymer additives effect. The prediction results are directly compared to the DNS data to assess the performance of the present model predictions.

Effect of the nonuniform grid on the two-dimensional transport equation was investigated in terms of theoretical analysis and finite difference method (FDM). The nonuniform grid having a typical structure of the numerical weather forecast model was incorporated in the vertical direction, while the uniform grid was used in the zonal direction. The staggered and non-staggered grid were placed in the vertical and zonal direction, respectively. Time stepping was performed with the third-order Runge Kutta scheme. An error analysis of the spatial discretization on the nonuniform grid was carried out, which indicated that the combined effect of the nonuniform grid and advection velocity produced either numerical diffusion or numerical adverse-diffusion. An analytic function is used for the quantitative evaluation of the errors associated with the discretized transport equation. Numerical experiments with the non-uniformity of vertical grid were found to support the analysis.

An algebraic model for turbulent heat fluxes is proposed on the basis of the elliptic blending equation. The algebraic model satisfies the temperature-pressure gradient correlation characteristics of near-wall region and the flow center region far away from the wall. That is, the turbulent heat flux conditions for both regions are connected by the solution of the elliptic blending equation. The predictions of turbulent heat transfer in a plane channel flow have been carried out with constant wall heat flux and constant wall temperature difference boundary conditions respectively. Also, the rotating channel flow with constant wall temperature difference is considered to test the applicability of the model. The prediction results show that the distributions of the turbulent heat fluxes and mean temperature are well captured by the present algebraic heat flux model.